IR 05000237/1991009
| ML17202V067 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 04/16/1991 |
| From: | Burgess B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17202V064 | List: |
| References | |
| 50-237-91-09, 50-237-91-9, 50-249-91-08, 50-249-91-8, NUDOCS 9104230383 | |
| Download: ML17202V067 (16) | |
Text
- .
U. S. NUCLEAR. REGULATORY COMMISSION
REGION III
Report No ~237/91009(DRP); 50-249/91008(DRP)
Docket No * 50-249
' '
'
Litensee:
Commonwealth Edi~on Company P. 0. Box 767 Chicago, IL 60690 License No Facility Name:
Dresden Nuclear Power Station, Units 2 and 3 Inspection,At:. Dresden Site, Morris, IL DPR-19* DPR-25
'
Inspection Conducted:
February 16, 1991 through April 2, 1991 Inspectors:
D. Hills M. Peck M. Kunowski J. Monninger R. Zuffa, Site Resident Engineer Illinois Department of Nuclear Safety ApproVed By: cgffe.~~
Reactor Projects Section 18 Inspection Summary Date * *
Inspection from February 16 through April 2, 1991(Report Nos. 50-237/9l009(DRP);
50~249/91008(DRP))
Areas Inspected:
Routine unannounced safety inspection by the resident inspectors, and an Illinois Department of Nuclear Safety resident engineer of licehsee action on previously identified.items; licensee event reports followup; operational safety; monthly.maintenance; monthly surveillance; training effectiveness; events; engineering and technical support; safety assessment and quality verificatic:rn; followup on TMI action items; Sys_tematic Evaluation Prpgram items; report review; and meeting * *..... **-
-
Results:
One cited violation was identified involving inadequate post-modification testing of the standby liquid control system (paragraph 2).
Four non-cited violati6ns were identified involving an inadequate 10 CFR 50.59 safety evaluation, (paragraph 2); a control rod mis-manipulation (paragraph 8);
a loss of torus to drywell differential pressure due to personnel error (paragraph 8); and a failure to perform expanded safety valve testing (paragraph 9).
One unresolved item was identified involving the application of safety/non-safety*electrical separation to a temporary alteration (paragraph 6).
9104230383 910418 PDR ADOCK 05000237 G
'
i,
-*
- e Plant Operations This lrea remained on a declining trend in regard to work ptacti~es. Thi$ was exhibited by a control rod withdrawal error, a loss of drywell to torus di-fferential pressure caused by inattention to detail *in implementing a procedure, and a possible out of servic~ error resulting in contamfnation of a 1arge portion of the turbine buildin Improvemerits wer~ noted in plant hou~ekeepin~.
Maintenance/surveillance This area remains under close scrutiny due to the negative trend.in the work pra~tices identified in recerit inspection report No significant problems were noted in the observed maintenance/surveillance activities delineated in this repor Engineering and Technical Support This area remained mixe The licensee's identification of the failure to obtain a Tech~ical Specification amendment for the turbine contrril valve sole~oid fast closure.swit~h modification indicated good application of experience from other CECo facilitie However, the failure to perform expanded safety valve sample testing due to a misinterpretation represented a non-conservative approach to code requirement In addition,
post-modifi cat ibn testing of the standby 1 iquid _control system indicated deficiencies in design and testing document *
Radiological Protection Performance in this area showed an improving trend compared to deficiencies noied during the recent Unit 2 refueling outag The licensee implemented a ne~ ~adiation work permit ~rogram which is continuing to be evaluated for effectiveness. *Recovery activities following a turbine building contamination event were goo Emergency Preparedness Performance in this area remained good with two unusu~l events properly classified and execute Security J
Performance in this area remained good in regard to no problems being noted in security program implementatio Safety Assessment and Quality Verification Performance in this area continued to improv Although it was too early to judge effectiveness, management's identification and corrective actions in regards to work practice problems appeared extensiv Nuclear Quality Programs methodology improvements continued and scram/engineered safety features actuation reduction activities were considered-proactiVe. *
.2 I
I I
I I
.. I
!
DETAILS 1~
Persons Contacted Commonwealth Edison Company
- E. Eenigenburg, Station Manager L. Ger.ner, Technical Superintenden *J. Kotowski, Production Superintendent
- E. *Mantel, Services Director D. Van Pelt, Assist~nt Superintendent - Maintenance J. Achterberg, Assistant Superintendent - Work Planning
- G. Smith, Assistant Superintende~t-Operations
- K. Peterman, Regulatory Assurance Supervisor
- M. Korchynsky, Operating_ Engineer
- B. Zank, Operating Engi~eer J. Williams, Operating Engineer R. Stobert, Operating Engineer
- T. Mohr, Operating Engineer
'*M. Strait, Technical Staff Supervisor L. Johnson, Q.C. Supervisor J. Mayer, Sta ti on Security Administrator. Mor~y, Chemistry Services Supervisor D. Saccomando, He~lth Physics Services Supervisor
- K. Kociuba, Quality Assurance Superintendent
- D. Lowenstein, Regulatory Assurance Analyst
- T. Gallaher, Quality Assurance Engineer
- B. Viehl, Nuclear Engineering Department, Supervisor
- Denotes thos~ attending the exit interview conducted on April 2, 1991, and at other times throughout the inspection perio The inspectors also talked ~ith and interviewed several other licensee employees, including members of the technical and engineering staffs, reactor and auxiliary operators,' shift engineers and foremen, electrical, mechanical and instrument maintenance personnel, and contract security personne.
Previously Identified Inspection Items (92701 and 92702)
(Closed) Unresolved Item (50-237/90017-06):
The NRC completed review of the standby liquid control (SLC) system suction piping anticipated transient without scram (ATWS) modification (M12-2(3)-84-119) design calculations and post modification tes The calculations indicated minimum net positive suction head (NPSH) was analytically indeterminate and relied on the performance of a post modification test to verify the desig The test specification, dated September 24, 1986, was inadequate such that the dual SLC pump pas~ modification test (performed on February 18, 1987 for Unit 2; and May 17, 1988 for Unit 3) failed to measure the critical parameters, including the vapor pressure of the test solution and pump volumetric efficienc The SLC post-modification test
-
.
was als6 iriadequate baied on the insufficient duration {approximately one minute) to reach hydraulic stability and for _the discharge pressure to be prop~rly throttled._ This was in comparison to the Unit 2 dual pump special test performed on February 15, 1991, in which approximately 120 seconds were _required to throttle-the discharge pressure of the running pumps to the requited 1275 psig __ *.
.
.
.
The failure to specify post modification :test acceptance criteria to verify the SLC ATWS NPSH design under the most adverse design conditions, and the failure to perform a two pump-post-modification flow test, of __
an adequate duration to demonstrate satisfactory inservice performance, -
are a violation of 10 CFR 50, Appendix B, Criterions *III and XI
(50-237/91009-0l(DRP))~
-
- -
-
. _
(Closed) Unresolved Item (50-237/90019-03):
Determine plant ~~ecific requirements and commitments regarding control and protection circuitry separatio On July 30, 1990, a Group V isolation occurred when a power supply fuse opened following the replacement 1 of a burned out light bulb in a control room low pressure coolant injection (LPCI) system manual isolation valve indicator. The LPCI and Group V isolation circuitry int~rface had existed since original plant construction. A review of the applicable codes indicated the electrical configuration was consistent with the original design bases. Additionally, a review of the Systematic Evaluation Program (SEP) did not identify any changes to original requirements related to* control _and -protection circuitry separation. *The inspectors have no other concerns in this are~.
.
-
-
(Closed) Unresolved (50-237/90027-05):
The NRG.completed the review of th~ licens~e 1 s corrective actions associated with the December 8, 1990, Unit 2 partial Group II isolation. *The corrective actions *included restoration of the control room panel wiring consistent with the plant drawing The inspectors randomly field verified the wiring diagrams associated with three main control room panels and identified several additional drawing deviation No safety significance existed with any of the additional deviation The licensee committed to correct the drawing deviations*, which is consistent with NRG/licensee agreements regarding this issue from a previous violation (50-237/86015-01; 50-249/86017-01).
The inspectors have no further concerns in this afe (Closed) Unresolved Item (50-237/90027-11):
In 1984, the licensee replaced the turbine control valves' (TCV) fast actihg solenoid valve *
limit switches with pressure switches (PS). These PSs provided input into the reactor protection system (RPS) to generate a reactor sc~am under load reject condition The PSs required periodic calibration as opposed to the previous simple on/off type limit switches. Technical_
Specjfication Basis stated calibration requirements were not included for these instrument channels since they were simple on/off switche Technical Specifications were not amended to reflect the modification and the periodic calibration was not performed. Although some of the setpoints drifted below the lower bounds of the setpoint calculation, the
-~*
.. ~-----~
Document Name:
~
DRESDEN 91009,008 Requestor 1 s. ID:
HAUSMAN Author's Name:
Burgess Document Comments:
- licensee verified the PS as-found response times were consistent with the accident analysis. This minimized the safety significance of the issu The.modification safety evaluation, performed November 4, 1981, incorrectly concluded a change in ~he Techni~al S~ecifications were not required. This was the result of an inadequate utilization of the _FSAR during the safety evaluation process. However, this issue was already encomp.assed'by the broad ranging correC:tive actions, delineated in a December 27, 1990 licensee letter to the NRC in response to a previous violation~ A licensee safety evaluation, perf6fmed_~ubsequent to the identification of the issue, conclud~d that a Technical Specification amendment was require The am.endment request was submitted January 23, 199 Section 11.2.3.2 of the Updated Final Safety Analysis Report (FSAR),
prior to revision 7, and corresp6nding portions of the original FSAR stated that the generator load rejection* scram was initiated by position switches on the control valve solenoid The licensee, without prior NRC approval, made a change in the facility as described in the safety-analysis report by changing the initiation of the generator load reject
~cram from position to pressure switche This involved a change iri Technical Specification scram instrumentation calibration requirements and, therefore, was a violation (50-237/91009-02(DRP)) of 10 CFR 50.59(a).
However, as this violation corresponded to the criteria for the exercise of discretion delineated in 10 CFR 2, Appendix C, Section V.G.-1, a Notice of Violation is not being issue (Open) Open Item (50-237/90027-14):
Perform sample inspection of SEP topic resolution An additional SEP item confirmed by the inspectors is listed in-paragraph 1 The open item will remain open pending completion of licensee confirmation of topic closures and completion of the NRC sample inspectio One cited and one non-cited violation and no deviations were identified in this are.
License Event Reports Followup (90712 and 92700)
Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specifications. _ (Closed) LER 237/90-003:
Drywell Air Sampling - Containment Leakag (Closed} LER 237/90-004:
Loc~l Leak Rate Testing
. (LLRT) - Additional Volumes Added to Type B & C (Due to Self Assessment).
- (Closed) LER 237/90-011:
Diesel* Generator 2/3 Unplanned Automatic Start Due to Procedure Deficienc d. *
(Closed) LER 237/90-013:
Unknown Reactor Building Ventilation Duct Sample Flow Due to Loose Moisture Separator Ja (Closed) LER 237/90-015:
Intermediate Range Monitor (IRM)
F~ll Scram Due to Inductive Noise Input to the IRM/Sourc~ Range Monitor (SRM) Power Supplie *
- (Closed) LER 237/90-016:
Standby Liquid Control (SLC) Piping Found in Violation_of FSAR Design Criteria Due to Management Deficienc (Closed) LER 237/90-017:
Reactor Scram on Intermediate Range
. Monitor Hi-Hi Due to Unknown Caus (Closed) LER 237/90-018:
Leakage Path Discovered During Primary Containment Integrated Leak Rate test (ILRT) Due to Management Deficienc.
(Closed) LER 237/90-019:
Electromatic Relief Valve Pressure Switch Outside Technical Specification Due to Instrument Setpoint Drif (Closed) LER 237/90-021:
Main Steam Safety Valves 2-203-4E through 4H Setpoints Found Outside Technical Specification Limits Due to Setpoint Drif (Closed) LER 237/90-022:
Unexpected Closure of Eleven Containment Isolation Valves During Surveillance Testing Due to Procedure Deficienc.
(Closed) LER 237/90-023:
Channel 118 11 Drywell Radiatfon Monitor Detector Found Inoperable Due to an Internal Detector Faul (Closed) LER 249/90-006:
Failure to Establish Appropriate Fire Inspections Due to Procedure Deficienc In addition to the foregoing, the inspector reviewed the licensee 1 s Deviation Reports (DVRs) generated during the inspection perio This was done in an effort to monitor the conditions related to plant or personnel performance, potential trends, et DVRs were also reviewed for initiation and disposition as required by the applicable procedures and the Quality Assurance (QA) manua No violations or deviations were identified other thah th6ii addr~ssed in this and other NRC inspection report.
Operational Safety Verification (71707)
During the inspection period the inspectors verified daily, and randomly during back shift and on weekends, that the facility was being operated in conformance with licensee and regulatory requirements and that the
- licensee's management control system_was effectively carrying out its
. responsibilities for safe operatio This was done on
~ sampling basis through routine direct observation of activiti~s and equipment, tours of the facility, interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions_
f6r op~ration (LCO) action requirements, corrective action, and review of facility record On a sam.pling basis the inspectors dafly verified proper control room*
staffing and access, operator behavior, and coordination of plant acti~ities with ongoing control roo~ operations; verified operator adherence with the latest revisions of procedures for ongoing activities; verified operation as required by Technical Specifications (TS); including compliance with LCOs, with emphasis on engineered safety features (ESF) and ESF electrical ~lignment and valve positions; monitored instrumentation recorder traces and du~l{cate channels for abnormalities; verified status of various lit annunciators for operator understanding, off-normal condition, and corrective actions being taken; examined nuclear instru~entation (NI) and. other protection channels for proper operability; reviewed radiation monitors and stack monitors for
- abnormal conditions; verified that onsite and offsite power was available as required; observed the.frequency of plant/control room visits by the station manager, sup~rfntendents, assistant superintendents, and other managers; and observed the Safety Parameter Display System (SPDS) for operabilit *
During t6urs of atcessible areas of the plant, the inspectors made note of general plant/equipment conditions, including control of activities in progress (maintenance/surveillance), observation of shift iurnovers, general safety items, et The specific areas observed were: Engineered Safety Features ( ESF) Systems Accessible portions of ESF systems and components were inspected to verify:
valve position for proper flow~ path; proper alignment of power supply breakers or fuses (if visible) for proper actuation on an initiati~g signal; proper removal of power from components if required by TS or FSAR; and the operability of support systems-essential to system actuation or performance through observation of inst~umentation and/or proper valve alignmen The inspectors also visually inspected components for leakage, proper lubrication, cooling water supply, et Radiati-0n Protection Controls The inspectors verified that workers were following health physics procedures for dosimetry, protective clothing, frisking, posting, etc., and randomly examined radiation protection instrumentation for use, operability, and calibratio The inspectors noted that the licensee instituted a computerized radiation work permit progra It was designed to allow better personnel accountability and easier manipulation and tracking of data for use in dose reduction activitie.
--c-I, I
I I
- c. Security.
Each week duririg routine ~ctfvities or tours, the inspector monitored the licensee's security program to ensure that observed actions.were being implemented according to their approved security pla The inspect6r noted that persons within the protected are displayed proper photo-identificatioh badges and those individuals requiring escorts were properly escorted.. The inspector also verified that checked vital areas were locked and alarme Additionallyr the inspector also verified that observed personnel and packages entering the protected area were searched by appropriate equipment or by han * H6useke~pi~~ i~d ~larit Cle~hliness The inspectors monitore~ the status of housekeeping and plant cleanliness for fire protection, protection of safety-related e~uipment from intrusion of foreign matter and general protection of equipment from hazard The inspectors noted that housekeeping had improved from the ~oor conditions observed during the recent Unit 2 refueling outag The inspectors* also monitored various records, such as tagouts, jumpers, shiftily logs and surveillances, daily orders, maintenance items, various chemistry and radiological sampling and analysis, third party review results, overtime *records, Nuclear Quality Programs (NQP) and/or Quality Control (QC) audit results, and postings* required per 10 CFR 19.1 *
.
.
No violations or deviations were identified in this are.
. Monthly Maintenance Observation (62703)
Station maintenance activities affecting the safety..::related systems. and components listed below were observed/reviewed to ascertain that they were conductad in accordance with approved procedures, regulatory guides, and industry codes or standards and in conformance with Technical Specification The following items were considered during this review:
the Limiting Conditions for Operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemente Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may affect system performanc *
- The following maintenance activities were observed and reviewed:
Unit 2 Source Range Monitor Troublesho~ting Unit 3 Low Pressu~~ Coolant Injection Pu~p 30 Overhaul The inspectors monitored the licen~ee 1 s w~rk in progress and verified that it was being performed in accordance with proper procedures, and approved work packages, that applicable draw*i ng updates were made and/or planried, and that operator training was conducted in a reasonable period of tim *
'
The inspectors reviewed proactive instrumentation hardware upgrades, performed during 1990, to address ~ecurring maintenance and procurement problems.. Most notable were a~ upgrade of the control room recorders, source range and intermediate range monitor cable and connectors, digital
- HPCI flow transmitters and squareroot converters, and the f~ll core display lightin These reflected instrument maintenance department aw_areness of operator concerns and problem No violations or deviations were id~ntifie.
Monthly Surveillance Observation (61726)
The inspectors observed surveillance testing required by Technical Specifications duning the inspection period and verified that testing was performed in accordance with adequate proced~res, that test instrumentation was calibrated, that LCOs were met; that removal and restoration of the affected components were accomplished, that results conformed with Technical Specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne The inspectors witnessed portions of the following test activities:
Unit 2 Fire System Yard Loop Monthly Inspection High Pressure Coolant Injection (HPCI) Pump Flow Tes Unit 3 Fire System Yard Loop Monthly Inspection Low Pressure Coolant Injection (LPCI) Full Flow Test Isolation Condenser High Flow Tes On February 8, 1991, the Unit 2 HPCI system was modified, under Temporary Alteration (TA) II-7-91, to incorporate measuring and test equipment (M&TE) for data acquisition during turbine cold start surveillance
testin The TA, which was left in place for a two month duration, provided dir.ect interface between the non-safety M&TE and the HPCI class lE electrical equipmen The licensee committed to the.NRC, by letter
- dated March 6, 1985, to incorporate the philosophy of IEEE-384 and.. *
Regulatory Guide 1.75 into all plant inodifii;:ations whenever practica IEEE-384 ~pecffied, in part, the electrical 'isolation criteria for the safety to non-safety interface for instrumentation and control circuit The TA 10 CFR 50.59 safety evaluation did not address the separation
r~quirements of IEEE-3S4 or incorporate the use of isolation devices at the interfac This issue is considered an unresolved item
- (50-237/91009-03(DRP)) pending further clarification on the applicability of both IEEE-384 and Regulatory Guide 1.75 in the current design and licensing basis of the plant and applicability to the HPCI TA.
. No violation~ or deviations were identified in this are ~
Training Effectiveness (41400 and 41701)
The effectiveness of training programs for licensed and non-licensed personnel was reviewed by the inspectors during the witnessing of the
- licensee's performance of routine surveillance, maintenance, and*
. operational activities and during the review of the licensee's resporise to.events which occurred during the inspection perio Personnel appeared to be knowledgeable of the tasks being performed, and nothing
- was observed which indicated any ineffectiveness of trainin No violations or deviations were identifie.
Events (93702) On February 19, 1991, while performing Dresden Operations -
. Surveillance (DOS) 300-2, "Control Rod Drive Stall -Flow Test for Units 2 and 3, 11 a control rod mis-manipulation occurred on Unit The procedure specified only testing of control rods at position 00 or position 4 Although both the full core rod position display and the four rod display indicated the selected rod, the Nuclear Station Operator (NSO) withdrew control rod s~a from position 08 to position 1 The NSO had instead focused attention on the control rod drive flow and pressure meter The NSO bypassed a rod block generated as a result of the mis-manipulation, believing it was related to the normal rod block monitor (RBM) nulling process and local power range monitor nois Following self-identification of the error, the NSO instinctively re-inserted the control rod without first notifying, and receiving permission from, the Qualified Nuclear Engineer (QNE).
This action was contrary to Dresden Operating Abnormal (DOA) 300-12,
"Mis-positioned Control Rod. 11 The safety significance of the event was minimal since no thermal limits were approache The cause of the event was a failure to ad~quately monitor, or respond to; all applicable indications and alarm The licensee restricted all power changes until the investigation was complete, tailgated the event with all incoming operating crews,
..
and reviewed DOA 300-12 wit~ all shift lic~nsed pe~sonnel *.
Administrative controls were instituted to ensure reprogramming of the rod worth minimizer (RWM) if an out of sequence was to remain greater than one shift, and to ensure no routine rod movement without art operable RWM, unles~ an independent verifier was present. A QNE was on site during the Unit.2 Cycle 13 startup testing during any'planned control rod motio In addition, the licensee planned to review possible enhancements to eliminate the need to bypass the RWM during scram testing, control*rod exercising,
. and stall flow testing and to increase the RBM setpoint such that the frequency of the intermittent steady-state rod b'locks could be reduce The failure of the NSO to adequately adhere_ to the r~quirements of DOS 300-2 and DOA 300-12 was considered to be a violation (50~237/91009-04(DRP)) of Technical Specification 6.2.A.3 ~nd 6.2. Recent violations did not involve inadequacies in indication and alarm monitoring, and therefoie, the inspectors considered this to be an isolated occurrence of minor safety significanc As appropriate corrective actions were also initiated, a Notice of Violation is not being issued in accordance with the exercise of discretion delineated in.10 CFR 2, Appendix C, Section On February.19, 1991, while pressurizing the cation tank during a sonic cleaning evolution, contaminated.water/resin w~s blown out of the cation tank fill funne Several minor personnel contaminations resulted and the licensee evacuated the turbine building. The development and implementation of the recovery was good with less than 0.5 pe~son-rem expended for the decontamination effec No intakes of radioactivity were recorde The cause of the event was the cation tank fill valve inappropriately placed in the open positio On February 18, 1991; a non-licensed operator had verified the valve was closed and recorded it closed on Equipment Outage Form II-21 In* subsequent interviews, the operator stated that the valve was closed as recorded on Form II-21 As no other evolutions required manipulation of the valve, the root cause could not be positively determine However, the licensee took extensive and comprehensive corrective actions and assumed that an error had occurred with the equipment outage. This.included: tailgating the event to operating crews involved in resin work; changing the procedure to ensure the valve was closed prior to pressurizing the tank;.and independent verification on all resin system out of service The licensee also planned to perform a review to determine if indep~ndent verific~tions should be included for additional non-safety system On February 19, 1991, while performing Dresden Operating Surveillance (DOS) 1600-1, Revision 15, "Quarterly Valve Timing,"
the one psid drywell to torus differential pressure, required by Technical Specification 3.7.A.7a, was lost on Unit 3 for 37 minute As this time was well within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> re-establishment Technical Specification action statement, minimal
safety significance was involve The event was caused by the NS0 1*s failure to turn to ~he next page in the procedure which prescribed the testing sequence, causing the valves to be cycled in an incorrect orde Poor human factors procedure format ~nd control board mimic deficiencies contributed to this erro The licensee instituted a temporary procedure change to ensure proper valve sequenc The licensee also planned to correlate the corresponding procedures with the checklists during the pro~edure upg~ade and to re-label the mimic~ Attention to detail was ~tressed to all personnel in ~he Febr.uary 26, 1991.meetings described in paragra,ph 1 The failure to adhere to the procedure is considered to be a violation of Technical Specification 6.2.A.7 (50-237/91008-0l(DRP)).
As the*
intent of the NSO was to fqllow the ~rocedure, the causes of this event was dissimilar to* other.recent procedural adherence violations. *This violation co~responded tb the crit~ria for exercise of discretion delineated in 10 CFR 2, Appendix C, Section V.G.1; the~eforei a Notice of Violation is not being issue * On March 4, 1991, the licensee identified excessive leakage through the recirculation sample inboard and outboard primary containment isolation valve A plant shutdown was initiated per Technical*
Specification 3.7.D.3 and an unusual event (UE) was declare Subsequent local leak rate testing indicated the Technical
Specification combined type Band C leakage limits were not exceeded.. On March 5, 1~91, wind speed and direction instr4ments tin.the meteorological tower.became inoperable due to heavy icing condition An UE was declared due to the loss of offsite dose assessment capabilit The UE was terminated when operability was restored to the instrument No cited and two non-cited violations and no deviations were identifie. *
En~ineering and Technical Support (92700)
Main steam safety valve as-found setpoirit testing, performed
- November 4, 1990, determined that all four valves*were outside of the plus or minus one percent criteria specified in Technical Spetification 4. The American Society of Mechanical Engineers (ASME) Boiler Pressure Vessel Code, Subsection XI, Section !WV 3513, (1977 Edition, through the summer of 1979 Addenda), prescribed expanded sample testing when any particular valve in a section failed to function properl Th~
~
inspectors identified that the licensee failed to perform***the additional tests required by the cod The failure was due to the inappropriate use of a plus three percent expansion acceptance criteria from the 1983 Edition (through the 1986 Addenda) of the Cod On March 9, 1991, the NRC verbally granted relief from the additional testing requirements contingent on the completion of a vessel overpressurization transient analysis at the plus three percent limi The analysis was completed by the fuel vendor on March 20, 199 Additionally, the licensee revised
.. **
Oresden Administrative Procedure (OAP) 11-21, "Inservice Inspection Program For Pumps and Valves, 11 to include a caution concerning the uses 6f acceptance criteria from later versions of the cdd The failure to perform, additional as-found testing, as required by the 1977 edition of the ASME Boiler and Pressure Code,Section XI, is considered a violation (50-237/91009-05(DRP)) of 10 CFR 50.55 However, as.this was considered to be an isolated occurrence and the appropriate corrective actions ~ere completed, a Notice of Violation is not being issued in accordance with 10 CFR 2, Appendix C, Section V.A~ The inspector has no further concerns in this are *
No cited and one non-cited violation and n~ deviations were identifie. * Safety Assessme.nt and Quality Verification (35502 and 40500) - The licensee took additional corrective action to stem the recent high rate of personnel error The inspecto~s ob~erved one of th~
plant manager presentations to all station personnel on February 26, 199 Events depicting poor work practices that occurred during and since the previous Unit 2 refueling outage were reviewe Procedural adherence, communications, self checking,
-attention to detail ~nd an inquisitive attittide were stresse These were also discussed in light of recent Quad Cities problem Departmental meetings were conducted on February 27, 1991 to facilitate the free exchange of concerns and ideas betwee manag~ment and th~ worker Licensee u~per management, including
- the Vice Pre~ident, BWR Operations, and the General Manage~, BWR, attended the meeting * The *inspectors noted recent changes in the Nuclear Quality Program (NQP) methodology to stress. performanc~ based audits and surveilla~ces. In addition, improvements in tracking and trending of deficiencies were institute The inspectors reviewed the Scram/Engineered Safety Features (S/ESF)
Actuation Reduction Progra The program provided a proactive approach to reduce future unplanned S/ESF actuations by applying a cost benefit evaluation and incorporation of hardware and programmatic change The inspectors reviewed several on-going plant modifications which originated as a result of the S/ESF progra The scram reduction committee also reviewed industry scram experienc.
Followup on TM! Action Items (2515/065-01)
By letter dated February 12, 1985, the NRC approved plans for modifications to improve reactor water level instrumentation in response to Generic Letter 84-23 "Reactor Water Level Instrumentation in BWRs.
This correspondence indicated these modifications close out the requirements for TM! Action Plan Item II.F. In a letter dated August 21, 1990, the licensee provided a revised schedule completion of
I
'I
the next refueling outag~ for Unit 2 and the next two ~~fueli.ng outages for Unit This item remains open pending cdmpletion of these
- modification No violations or deviations were identifie.
Systematic Evaluation Program Items (92701). *
NUREG 1403, 11Safety Evaluati.on Report Related to the Full-term Operating License for Dresden Nuclear Power Station, 11 Table 2.1 identified SEP Integrated Plant Safety Assessment Report (IPSAR) topic resolutions* to be confirmed by the NRC Region III offic The following item in that report was confirmed as closed by the inspectors:
(Item (15))
Topic III-6/4.9.3 -
NUREG~0823, Supplement No. indicated that this issue would be resolved on a
~ene~ic basis within.the framework of the NRC 1i
.
resolution of unresolved safety issue (USI) A-46 and, therefore, was considered-closed with regard to the SE.
No violations*or deviations were identified in this are.
Report Review (90713)
During the inspecti6n peridd, the inspector reviewed the licensee's Monthly Operating Report for January 1991.. The inspector confirmed that the information provided met the requirements of Technical Specification 6.6.A.5 and Regulatory Guide 1.16. The inspector also reviewed the Dresden Nuclear Power Stati-0n Monthly Plant Status Report for February 199 No violations or deviations were identified.*
1 Meetings and Other Activities On March 12, 1991, an on-site management meeting was conducted between the NRC and the licensee to discuss the status of the licensee's maintenance improvement program and the recent negative trend in events caused by personnel error The inspectors concluded good progress had been made in maintenance improvement but considerable activity remained for all actions to be fully implemente The licensee characterized the negative trend in personnel errors as a temporary setback and delineated extensive corrective actions to address the concer.
Violations For Which A 11Notice of Violation" Wi1"1 Not Be Issued The NRC uses the Notice of Violation as a standard method for formalizing the existence of a violation of a legally binding requiremen However, because the NRC wants to encourage and support licensee's initiatives for self-identification and correction of problems, the NRC will not generally issue a Notice of Violation for a violation that meets the requirements set forth in 10 CFR 2, Appendix C, Section Violations of regulatory requirements identified during the inspection for which a
I
. '
j Notice of Violation will not be issued are discussed in paragraphs 2, 8.a, 8.c and.
Unresolved Items-Unresolved items are matters about which more information is required in order to ascertain whether it is an acceptable item, an cipen item, a deviation or a violatio An unresolved item disclosed during this inspection is discussed in paragraph 6~.
1 Exit Interview The inspectors met with licensee representatives (denoted in paragraph 1)
during the inspection period and at the conclusion of the inspection period on April 2; 199 The inspectors summarized the scope and results of the inspection and discussed the likely content of this inspection repor The licensee acknowledged the information and did not indicate that any of the information disclosed during the inspection could be considered proprietary in natur